[Federal Register Volume 67, Number 199 (Tuesday, October 15, 2002)]
[Notices]
[Pages 63687-63706]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-25990]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, September 20, 2002, through October 3,
2002. The last biweekly notice was published on October 1, 2002 (67 FR
61674).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By November 14, 2002, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a
[[Page 63688]]
petition for leave to intervene shall be filed in accordance with the
Commission's ``Rules of Practice for Domestic Licensing Proceedings''
in 10 CFR Part 2. Interested persons should consult a current copy of
10 CFR 2.714,\1\ which is available at the Commission's PDR, located at
One White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April
29, 2002.
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland, by the above date. Because of continuing
disruptions in delivery of mail to United States Government offices, it
is requested that petitions for leave to intervene and requests for
hearing be transmitted to the Secretary of the Commission either by
means of facsimile transmission to 301-415-1101 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and because of continuing disruptions in delivery of mail to United
States Government offices, it is requested that copies be transmitted
either by means of facsimile transmission to 301-415-3725 or by e-mail
to [email protected]. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland. Publicly available records
will be accessible from the Agencywide Documents Access and Management
System's (ADAMS) Public Electronic Reading Room on the Internet at the
NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not
have access to ADAMS or if there are problems in accessing the
documents located in ADAMS, contact the NRC PDR Reference staff at 1-
800-397-4209, 304-415-4737 or by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: August 28, 2002.
Description of amendments request: The proposed change will revise
the expiration date of the facility operating licenses for Palo Verde
Nuclear Generating Station Units 1, 2, and 3, to recapture low-power
testing time.
Basis for proposed no significant hazards consideration
determination:
[[Page 63689]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Response: No.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because they do not involve a change to design configuration or
operation of the facilities. In addition, each PVNGS [Palo Verde
Nuclear Generating Station] unit was designed and constructed to
ensure a 40-year service life. Design features were incorporated
that provide for inspectability of structures, systems and
components during the 40-year service life. Surveillance,
inspectability and maintenance practices which have been implemented
in accordance with the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code and the unit Technical
Specifications provide assurance that any degradation in plant
safety-related equipment will be identified and corrected to provide
continued safe operation of each unit throughout the duration of the
applicable facility operating license.
The largest recapture period requested by the proposed amendment
requests is 8 months (Unit 3). This recapture period represents less
than 1.7% of the 40-year service life of the respective unit, and is
insignificant from an aging effects perspective. Therefore, the
proposed amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Response: No.
The proposed amendments would revise the expiration of each
facility operating license such that the expiration of each facility
operating license is based upon issuance of the respective FPOL
[full power operating license] and not upon issuance of the
respective LPOL [low power operating license]. No physical changes
are being made to the design features or operation of the
facilities. Therefore, the proposed amendments do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Response: No.
The proposed amendments would revise the expiration of each
facility operating license such that the expiration of each facility
operating license is based upon issuance of the respective FPOL and
not upon issuance of the respective LPOL. No physical changes are
being made to the design features or operation of the facilities.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary and the containment structure) to limit the
radiological dose to the public and control room operators in the
event of an accident. The proposed amendments to the facility
operating licenses are administrative in nature and have no impact
on the margin of safety and robustness provided in the design and
construction of the facilities. In addition, the proposed amendments
will not relax any of the criteria used to establish safety limits,
nor will the proposed amendments relax safety system settings or
limiting conditions of operation as defined in the Technical
Specifications. Therefore, the proposed amendments do not result in
a significant reduction in the margin of safety.
Based on the above information, APS [Arizona Public Service
Company] concludes that the proposed amendments present no
significant hazards consideration under the standards set forth in
10 CFR 50.92(c), and, accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: June 11, 2002.
Description of amendments request: The proposed amendment revises
Technical Specification (TS) 3.7.11, Spent Fuel Pool Exhaust
Ventilation System, for Units 1 and 2 to redefine the applicability of
the TS to limit the types of fuel assemblies to which it applies. This
proposed amendment revises TS 3.7.11 to not require the ventilation be
operable or in operation for the movement of fuel assemblies with an
appropriate amount of decay time. An evaluation has determined that 32
days is adequate time to allow for sufficient radioactive decay of
short lived isotopes resulting in no increase in offsite dose if the
ventilation system were not operable. This change is consistent with
changes previously approved for the Improved Standard Technical
Specifications as described in Technical Specification Task Force--51.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The system affected by this proposed amendment is the spent fuel
pool exhaust ventilation system (SFPEVS). This system mitigates the
consequences of a Fuel Handling Incident (FHI) by filtering
radioactive iodine from the air above the spent fuel pool prior to
that air being exhausted to the environment. This limits the offsite
dose possible from a[n] FHI. This proposed amendment revises the
Technical Specification applicability for the SFPEVS by defining
when the ventilation system is required to limit offsite dose due to
a[n] FHI. Because this system is used for the mitigation of an
accident, it is not an accident initiator. Therefore, the
probability of an accident previously evaluated is not increased.
The only design basis accident originating in the spent fuel
pool is the FHI. This accident is evaluated in the Updated Final
Safety Analysis Report. The analysis assumed credit for the
filtration system. However, a more recent evaluation shows that 32
days after a fuel assembly has been removed from the critical
reactor core, adequate radioactive decay has occurred which
compensates for the filtration of the ventilation system. Thus, no
increase in offsite dose occurs under these conditions. Therefore,
the consequences of an accident previously evaluated have not
increased.
Therefore, the probability or consequences of an accident
previously evaluated have not significantly increased.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The SFPEVS is not being altered by this amendment request. No
changes are made in the way in which the SFPEVS is operated or in
the way fuel is moved in the spent fuel pool. The only change made
would allow some irradiated fuel assemblies to be moved in the spent
fuel pool without requiring the operation of the ventilation system.
Since no changes are being made to the operation of the SFPEVS when
it is needed for offsite dose control and the SFPEVS is a[n]
accident mitigating system only, changes in when this system is
needed to operate cannot create a new type of accident.
Therefore, the possibility of a new or different [kind] of
accident from any previously evaluated is not created.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety provided by the SFPEVS is to limit offsite
dose due to a[n] FHI to the limits described in the Updated Final
Safety Analysis Report. The evaluation performed indicates that
radioactive decay can compensate for the filtration system. Thirty-
two days after fuel occupied a critical reactor core, enough
radioactive decay has occurred that the offsite dose from a[n] FHI
assuming no filtration is the same as the dose determined in the
Updated Final Safety Analysis Report. Therefore, no reduction in the
margin of safety has occurred because the
[[Page 63690]]
offsite dose is the same as the previously approved dose limits.
Therefore, the proposed changes do not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendment request: July 17, 2002.
Description of amendment request: The proposed amendment would
allow the installation of up to four lead fuel assemblies (LFAs)
Manufactured by Westinghouse Electric Company (Westinghouse) into the
Unit 2 Cycles 15 and 16 cores. Currently, Technical Specification (TS)
4.2.1, Fuel Assemblies, only allows fuel that is clad with either
zircaloy or ZIRLO. The Westinghouse LFAs utilizes advance zirconium-
based material for cladding. In addition, the statements currently in
TS 4.2.1 concerning the lead test assemblies that were allowed to be
inserted for Unit 1 Cycles 13, 14, and 15 will be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies,
states that fuel rods are clad with either zircaloy or ZIRLO. This
reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR Part
50, Appendix K, which also restricts fuel rod cladding materials to
zircaloy or ZIRLO. Calvert Cliffs Nuclear Power Plant, Inc. proposes
to insert up to four Westinghouse fuel assemblies into Calvert
Cliffs Unit 2 that have some fuel rods clad in zirconium alloys that
do not meet the definition of zircaloy or ZIRLO. An exemption to the
regulations has also been requested to allow these fuel assemblies
to be inserted into Unit 2. The proposed change to the Calvert
Cliffs Technical Specifications will allow the use of cladding
materials that are not zircaloy or ZIRLO for two fuel cycles once
the exemption is approved. To obtain approval of new cladding
materials, 10 CFR 50.12 requires that the applicant show that the
proposed exemption is authorized by law, is consistent with common
defense and security, will not present an undue risk to the public
health and safety, and is accompanied by special circumstances. The
proposed change to the Technical Specification is effective only as
long as the exemption is effective. In addition, the statements
concerning the exemption for Unit 1 Cycles 13, 14, and 15 have been
deleted, since Unit 1 Cycle 15 is completed, and therefore the
exemption has expired. The addition of what will be an approved
temporary exemption for Unit 2 and the deletion of an expired
exception to Technical Specification 4.2.1 does not change the
probability or consequences of an accident previously evaluated.
Supporting analyses indicate that since the lead fuel assemblies
(LFAs) will be placed in non-limiting locations, the placement
scheme and the similarity of the advanced alloy to ZIRLO will assure
that the behavior of the fuel rods with this alloy are bounded by
the fuel performance and safety analyses performed for the ZIRLO
clad fuel rods in the Unit 2 Core. Therefore, the addition of these
advanced claddings does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change does not add any new equipment, modify any
interfaces with existing equipment, change equipment's function, or
change the method of operating the equipment. The proposed change
does not affect normal plant operations or configuration. Since the
proposed change does not change the design, configuration, or
operation, it could not become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different [kind] of accident from any previously
evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The margin of safety for the fuel cladding is to prevent the
release of fission products. Supporting analyses indicate that since
the LFAs will be placed in non-limiting locations, the placement
scheme and the similarity of the advanced alloy to ZIRLO will assure
that the behavior of the fuel rods with these alloys are bounded by
the fuel performance and safety analyses performed for the ZIRLO
clad fuel rods in the Unit 2 cores. Therefore, the addition of the
advanced cladding does not involve a significant reduction in the
margin of safety.
The proposed change will add an approved temporary exemption to
the Unit 2 Technical Specifications allowing the installation of up
to four Westinghouse LFAs. The assemblies use the advanced cladding
materials that are not specifically permitted by existing
regulations or Calvert Cliffs' Technical Specifications. A temporary
exemption to allow the installation of these assemblies has been
requested. The addition of an approved temporary exemption to
Technical Specification 4.2.1 is simply intended to allow the
installation of the LFAs under the provisions of the temporary
exemption. The license amendment is effective only as long as the
exemption is effective. This amendment does not change the margin of
safety since it only adds a reference to an approved, temporary
exemption to the Technical Specifications.
In addition, the words concerning the exemption for Unit 1
Cycles 13, 14, and 15 will be deleted since Unit 1 Cycle 15 is
completed, and therefore, the exemption has expired. This change
does not change the margin of safety since it only deletes a
reference to an expired exemption to the Technical Specifications.
Therefore, the proposed change does not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 6, 2002.
Description of amendments request: The proposed amendment would
revise Technical Specification (TS) 3.9, Refueling Operations, to
incorporate two changes previously approved in NUREG-1432, Revision 2,
``Combustion Engineering Improved Standard Technical Specifications''
dated April 2001. One change would add a note to Limiting Condition for
Operation 3.9.3 allowing penetration flow path(s) that have direct
access from containment atmosphere to the outside atmosphere to be
unisolated under administrative control. The other change would replace
the requirement in TSs 3.9.4 and 3.9.5 to ``[c]lose all containment
penetrations providing direct access from the containment atmosphere to
outside atmosphere'' with a set of more detailed and less restrictive
requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR part 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 63691]]
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Closing the containment penetrations is considered to be a
mitigator of the radiological consequences of a fuel handling
incident and a loss of SDC [Shutdown Cooling], not an initiator.
Therefore, allowing containment penetration flow paths to be
unisolated and the containment purge valves to be opened during
these outage activities does not involve a significant increase in
the probability of an accident previously evaluated.
The consequence of a fuel handling incident is the release of
radioactivity from Containment. The impact of the proposed change to
the calculated offsite dose resulting from a fuel handling incident
has been evaluated and determined to be acceptable. The fuel
handling incident analysis assumes no containment closure. The
amount of radioactivity that could be released as a result of the
proposed change is bounded by the current analysis of record.
Therefore, having containment penetration flow paths unisolated
during core alterations and fuel handling does not involve an
increase in the consequences of an accident previously evaluated.
The consequences of a loss of SDC is the potential for release
of radioactivity to the atmosphere outside Containment. Closing
containment penetrations is a mitigator of that consequence.
Administrative controls will be put in place to ensure that in an
emergency containment closure can be quickly achieved. The
containment purge system isolation valves are closed automatically
on a containment high radiation signal and can be shut by remote
manual operation. Therefore, the proposed changes do not involve a
significant increase in the consequences of a loss of SDC.
Therefore, the proposed Technical Specification changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
This requested change does not involve a significant change in
the operation of the plant and no new accident initiation mechanism
is created by the proposed changes. Closing containment penetrations
is considered to be a mitigator of the radiological consequences of
any accident in the Containment, not an initiator. The containment
penetration flow paths are currently opened and closed during the
course of an outage. The proposed changes allow them to remain open
during a period when they are currently required to be closed.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety for containment closure during core
alteration/fuel handling is based on the amount of offsite dose
resulting from a fuel handling incident. An offsite dose calculation
previously approved by the Nuclear Regulatory Commission for a fuel
handling incident assumes no containment closure, and any activity
released from the Containment is unfiltered. The analysis will apply
to the containment penetration flow paths that could be opened under
administrative controls and therefore, does not involve a
significant reduction in the margin of safety.
The margin of safety for containment closure in the case of a
loss of SDC is twofold: (1) The time required to close the
Containment to prevent a radioactive release to the atmosphere
outside Containment if SDC is lost; and (2) the ability to retain
the pressure generated by boiling of reactor coolant as a result of
a loss of SDC.
Currently the Technical Specifications are vague and overly
restrictive concerning the requirement for containment closure when
SDC is lost. The proposed change eliminates unclear requirements and
provides a clear way to establish containment closure that meets the
Bases description for the Action, which is to prevent fission
products from being released from the Containment during a loss of
SDC incident. The containment purge isolation valves close rapidly
on a high radiation signal or are closed by remote manual operation.
The proposed changes do not increase the possibility of a release of
radiation following a loss of SDC incident.
Therefore, the ability to provide containment closure is
maintained and the margin of safety is not significantly reduce[d]
by this proposed activity.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: August 28, 2002.
Description of amendment request: The amendment, proposed by
Carolina Power & Light Company to the Harris Nuclear Plant (HNP)
Technical Specifications (TS), revises TS 6.9.1.6.2 to add analytical
methodology references, which are used to determine core operating
limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved by the NRC for use in licensing
applications. Plant structures, systems, and components will not be
operated in a different manner as a result of these proposed changes
and no physical modifications to equipment are involved. Adding
these references to the Core Operating Limits Report section of
Technical Specifications does not increase the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved by the NRC for use in licensing
applications. Plant structures, systems, and components will not be
operated in a different manner as a result of these changes and no
physical modifications to equipment are involved. Adding these
references to the Core Operating Limits Report section of Technical
Specifications does not create the possibility of a new or different
type of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes incorporate additional references to
methodologies used to evaluate core operating limits. These
methodologies have been approved by the NRC for use in licensing
applications. Plant structures, systems, and components will not be
operated in a different manner as a result of these changes and no
physical modifications to equipment are involved. Adding these
references to the Core Operating Limits Report section of Technical
Specifications does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe
[[Page 63692]]
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: August 7, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) related to safety system
settings. Specifically, the proposed changes would revise: (1) TS 1.0
``Definitions;'' (2) TS 2.2.1 ``Limiting Safety System Settings--
Reactor Trip System Instrumentation Setpoints;'' (3) TS 3.3.1 ``Reactor
Trip System Instrumentation;'' (4) TS 3.3.2 ``Engineered Safety
Features Actuation System Instrumentation;'' (5) TS 3.7.7 ``Control
Room Emergency Ventilation System;'' (6) TS 3.8.3.1 ``Onsite Power
Distribution--Operating.'' In addition, the appropriate TS Bases would
be revised to conform with the proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes associated with the operability
requirements, surveillance requirements and allowed outage times
will improve usability of the facility Technical Specifications. The
proposed changes will clearly reflect the existing plant design for
the Reactor Trip System (RTS), Engineered Safety Features Actuation
System (ESFAS), Control Room, Emergency Ventilation System, and
Electrical Power Systems Instrumentation. The proposed changes will
also provide consistency within the individual technical
specifications tables (e.g. Table 2.2-1, Table 3.3-1, and Table 4.3-
1). In addition, there are no hardware changes associated with the
proposed changes. Therefore, these systems will continue to perform
within the bounds of the previously performed accident analyses.
The proposed changes to the operability requirements will not
affect the instrumentation's ability to mitigate the design basis
accidents. The proposed allowed outage times (i.e. the required
action times) are reasonable and consistent with industry guidelines
to ensure the affected instrumentation will be restored in a timely
manner and provide consistency with the existing plant design. The
design basis accidents will remain the same postulated events
described in the Millstone Unit No. 3 Final Safety Analysis Report
(FSAR), and the consequences of these events will not be affected.
Therefore, the proposed changes will not increase the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. The proposed changes do not alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated. The proposed
changes do not introduce any new failure modes. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any accident analysis assumption. The
proposed changes do not decrease the scope of equipment currently
required to be operable or subject to surveillance testing, nor do
the proposed changes affect any instrument setpoints or equipment
safety functions. The effectiveness of Technical Specifications will
be maintained since the changes will not alter the operation of any
component or system, nor will the proposed changes affect any safety
limits or safety system settings which are credited in a facility
accident analysis. Therefore, there is no reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: James W. Andersen, Acting.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 12, 2002.
Description of amendment request: The proposed amendments would
temporarily revise Technical Specification (TS) 3.5.2, ``Emergency Core
Cooling System (ECCS);'' TS 3.6.6, ``Containment Spray System;'' TS
3.7.5, ``Auxiliary Feedwater (AFW) System;'' TS 3.7.7, ``Component
Cooling Water (CCW) System;'' TS 3.7.8, ``Nuclear Service Water System
(NSWS);'' and TS 3.8.1, ``AC Sources--Operating'' for Catawba Nuclear
Station, Units 1 and 2. The proposed TS changes will allow the ``A''
NSWS header for each unit to be taken out of service for 7 days for
pipe replacement. This pipe replacement is scheduled to occur when
Units 1 and 2 are at power operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Catawba is currently pursuing a project to replace a portion of
the `A' train of the nuclear service water system (NSWS) piping for
both units. This is necessary to maintain the long-term reliability
of the NSWS. This project represents a challenge in that it is not
possible to isolate, drain, replace, restore and test the NSWS
during the current TS action time frame. The purpose of this
submittal is to request a temporary change to the existing TS for
the systems affected during the project. This will permit an orderly
and efficient project implementation during power operation on both
units. The specific change is to extend the TS required action time
from 72 hours to 168 hours.
The following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendment
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
First Standard
The pipe replacement project for the NSWS and proposed TS
changes have been evaluated to assess their impact on normal
operation of the systems affected and to ensure that the design
basis safety functions are preserved. During the pipe replacement
the other NSWS train will be operable and no major maintenance or
testing will be done on the operable train. The operable train will
be protected to help ensure it would be available if called upon.
This pipe replacement project will enhance the long term
structural integrity in the NSWS system. This will ensure that the
`A' NSWS header maintains its flow margin to ensure its ability to
comply with design basis requirements and increase the overall
reliability for many years.
The increased NSWS train unavailability as a result of the
implementation of this amendment does involve a one time increase in
the probability or consequences of an accident previously evaluated
during the time frame the NSWS header is out of service for pipe
replacement. Considering this small time frame for the `A' NSWS
train outage with the increased reliability and the decrease in
unavailability of the NSWS system in the future because of this
project, the overall probability or consequences of an accident
previously evaluated will decrease.
An evaluation was performed utilizing PRA [probabilistic risk
analysis] for
[[Page 63693]]
extending the NSWS TS time limit from 72 hours to 168 hours. The
[CDF] core damage frequency contribution from the proposed outage
extension is judged to be acceptable for a one-time, or rare,
evolution. Considering the change in CDF associated with the outage
extension in the framework of an average over a five-year period,
the average annual contribution is considered a low-to-moderate
increase in the CDF for consideration of permanent changes to the
licensing basis.
Therefore, because this is a temporary and not a permanent
change, the time averaged risk increase is acceptable. The increase
in the overall reliability of the NSWS along with the decreased
unavailability in the future because of the pipe replacement project
will result in an overall increase in the safety of both Catawba
units. Therefore, the consequences of an accident previously
evaluated remains unaffected and there will be minimal impact on any
accident consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed temporary TS changes do not
affect the basic operation of the ECCS [emergency core cooling
system], containment spray system, NSWS, AFW [auxiliary feedwater],
CCW [component cooling water], or EDG [emergency diesel generator]
systems. The only change is increasing the required action time
frame from 72 hours to 168 hours (ECCS, containment spray system,
NSWS, AFW, CCW, and EDG). During the project, contingency measures
will be in place to provide additional assurance that the affected
systems will be able to complete their design functions. No new
accident causal mechanisms are created as a result of NRC approval
of this amendment request. No changes are being made to the plant,
which will introduce any new accident causal mechanisms.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed temporary TS amendment. During the `A' NSWS train outage,
the affected systems will still be capable of performing their
required functions and contingency measures will be in place to
provide additional assurance that the affected systems will be
maintained in a condition to be able to complete their design
functions. No safety margins will be impacted.
The probabilistic risk analysis conducted for this proposed
amendment demonstrated that the CD[F] associated with the outage
extension is judged to be acceptable for a one-time or rare
evolution. Therefore, there is not a significant reduction in the
margin of safety.
Based upon the preceding discussion, Duke Energy has concluded
that the proposed amendment for a temporary one time TS change does
not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: September 3, 2002.
Description of amendment request: The proposed changes to the
Columbia Generating Station Technical Specifications (TSs) are to: (1)
Add depleted uranium to the fuel assembly composition description in TS
4.2.1, (2) revise TS 5.6.5.b to incorporate references to the
analytical methods used to determine core operating limits and remove
those that are no longer used, and (3) format the revised references as
described in Industry/Technical Specification Task Force (TSTF)
Traveler, TSTF-363, ``Revised Topical Report References in ITS
[Improved Technical Specifications] 5.6.5, COLR [Core Operating Limits
Report].''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Assembly and core designs employing depleted uranium are
employed in other reactors and are within the FRA-ANP fuel design
methods and experience base. There will be no change to the
composition of the fuel pellets (i.e., UO2) containing
the depleted uranium except for a slight decrease in the amount of
U235. Therefore the use of depleted uranium in the fuel
rods does not affect the mechanical performance of the rods. Flux
profile measurements performed on these core designs correlate with
calculated values in a manner consistent with fuel assembly designs
that do not include depleted uranium.
Core operating limits are established to support Technical
Specification 3.2, Power Distribution, requirements which ensure
that fuel design limits are not exceeded during any conditions of
normal operation or in the event of any Anticipated Operational
Occurrence (AOO). The methods used to determine the core operating
limits for each operating cycle are based on methods previously
found acceptable by the NRC and listed in TS section 5.6.5.b. A
change to TS section 5.6.5.b is requested to include the FRA-ANP
methods in the list of approved methods applicable to Columbia
Generating Station. Application of these approved methods will
continue to ensure that acceptable operating limits are established
to protect the fuel cladding integrity during normal operation and
AOOs.
The requested Technical Specification changes do not involve any
plant modifications or operational changes that could affect system
reliability, performance, or possibility of operator error. The
requested changes do not affect any postulated accident precursors,
do not affect any accident mitigation systems, and do not introduce
any new accident initiation mechanisms.
Therefore, these changes do not increase the probability or
consequences of any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Assembly and core designs employing depleted uranium are within
the capability of the NRC-approved FRA-ANP fuel design methods.
There will be no change to the composition of the fuel pellets
(i.e., UO2) containing the depleted uranium except for a
slight decrease in the amount of U235. Therefore the use
of depleted uranium in the fuel rods does not affect the mechanical
performance of the rods.
Changes to the methodologies listed in the TS are
administrative. The proposed changes do not involve any new modes of
operation, any changes to setpoints, or any plant modifications. The
core operating limits will continue to be developed using NRC-
approved methods that account for the mixed fuel core design. The
proposed methods do not result in any new precursors to an accident.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Assembly and core designs employing depleted uranium are within
the capability of the NRC-approved FRA-ANP fuel design methods.
There will be no change to the composition of the fuel pellets
(i.e., UO2) containing the depleted uranium except for a
slight decrease in the amount of U235. Therefore the use
of depleted uranium in the fuel rods does not affect the mechanical
performance of the rods.
The core operating limits will continue to be determined using
methodologies that have been approved by the NRC.
On this basis, the implementation of the changes does not
involve a significant reduction in margin of safety.
[[Page 63694]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: September 19, 2002.
Description of amendment request: The proposed amendments would add
a new analytical method to Technical Specifications (TS) Section 5.6.5,
``Core Operating Limits Report (COLR).'' The proposed change supports
the core design efforts currently in process for the upcoming Unit 2
refueling outage scheduled to begin in January 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to LaSalle County Station, Unit 1 and Unit 2
Technical Specifications (TS), involves reference to a new fuel
analytical method in TS Section 5.6.5, ``Core Operating Limits
Report (COLR).'' This code package supports the methodology
currently being used by Framatome-ANP in the reload design and
analysis process.
The proposed change to TS Section 5.6.5 will add to the list of
methods used to determine the core operating limits, the fuel
analytical method that supports design of the LaSalle County Station
Unit 2 Cycle 10 reload that is currently scheduled to startup on
February 5, 2003. The addition of the approved method to TS Section
5.6.5 has no effect on any accident initiator or precursor
previously evaluated and does not change the manner in which the
core is operated. The NRC approved method has been reviewed to
ensure that the output accurately models predicted core behavior,
has no affect on the type or amount of radiation released, and has
no affect on predicted offsite doses in the event of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to TS Section 5.6.5 does not affect the
performance of any LaSalle County Station structure, system, or
component credited with mitigating any accident previously
evaluated. The use of a new analytical method, which has been
reviewed and approved by the NRC for the design of a core reload,
will not affect the control parameters governing unit operation or
the response of plant equipment to transient conditions. The
proposed change does not introduce any new modes of system operation
or failure mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to TS Section 5.6.5 adds the current
analytical method for design and analysis of core reloads to the
list of methods used to determine the core operating limits. The NRC
has approved for use by licensees the analytical method being added.
The proposed change does not modify the safety limits or setpoints
at which protective actions are initiated, and does not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, Exelon Generation Company concludes that the
proposed amendment presents a no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and accordingly, a
finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: August 7, 2002.
Description of amendment request: The proposed amendments would:
(1) Revise the surveillance frequency for air or smoke flow testing of
containment spray nozzles, as specified in surveillance requirements
(SRs) 4.6.2.1.d and 4.6.2.2.f, from once per 10 years to following
maintenance which results in the potential for nozzle blockage and
allows the use of a visual examination in lieu of an air or smoke flow
test; (2) eliminate the SR 4.6.2.2.e.3 criteria for the river water
flow rate through the Recirculation Spray System heat exchangers; and
(3) make minor clarifying changes to the text in Technical
Specification 3.3.1.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed changes to the containment spray system nozzle
surveillance frequency, the manner in which the nozzles are verified
to be unobstructed, and the elimination of the associated
Recirculation Spray System (RSS) flow rate requirement does not
introduce an initiator of any design basis accident or event. The
proposed changes do not adversely affect accident initiators or
precursors nor alter the configuration of the facility or the manner
in which the plant is maintained. The river/service water system
monitoring program ensures that the river/service water flow through
the RSS heat exchangers will be maintained. The proposed changes to
provide alternate wording for the P-13 function in the Reactor
Protection System solely for clarification of the current criteria
does not adversely affect accident initiators or precursors. Thus,
the proposed changes do not involve a significant increase in the
probability of an accident previously evaluated.
The proposed changes do not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits. Introduction of foreign
materials into the containment spray system from the exterior is
unlikely due to the location of the spray headers, the passive
nature of the nozzles, station foreign material controls, and the
fact that the containment spray headers are maintained dry above the
water level maintained in the Recirculation Water Storage Tank which
inhibits active degradation mechanisms such as corrosion. The
proposed amendment to eliminate the associated RSS flow rate
requirements and the text clarification for the P-13 function do not
introduce an initiator of any design basis accident or event. The
proposed changes are consistent with the safety analysis assumptions
and resultant consequences. Accident analyses potentially affected
by the proposed change have been reviewed and none are adversely
affected. Thus, the proposed change does not involve a significant
increase in the consequences of an accident previously evaluated.
[[Page 63695]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes to the containment spray nozzle
surveillance frequency, the manner in which the nozzles are verified
to be unobstructed, the elimination of the associated RSS flow rate
requirement, and the text clarifications for the P-13 function do
not involve any physical alteration of the plant (i.e., no new or
different type of equipment will be installed), subsequently no new
or different failure modes or limiting single failures are created.
The plant will not be operated in a different manner due to the
proposed change. All SSCs will continue to function as currently
designed. Thus, the proposed change does not create any new or
different accident scenarios.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes to the containment spray system nozzle
surveillance frequency, the manner in which the nozzles are verified
to be unobstructed, the elimination of the associated RSS flow rate
requirement, and the text clarifications for the P-13 function do
not involve revisions to any safety limits or safety system settings
that would adversely impact plant safety. No current setpoints are
altered by this change. The proposed amendment does not alter the
functional capabilities assumed in a safety analysis for any SSCs
important to the mitigation and control of design bases accident
conditions within the facility. The river/service water system
monitoring program ensures that the river/service water flow through
the RSS heat exchangers will be maintained.
All of the applicable acceptance criteria for each of the
analyses affected by the proposed change continue to be met. The
conclusions of the [Updated Final Safety Analysis Report] remain
valid. Thus, since the operating parameters and system performance
will remain within design requirements and safety analysis, safety
margin is maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 23, 2002.
Description of amendment requests: The proposed amendments would
revise Facility Operating Licenses (OLs) DPR-58 and DPR-74, for Unit 1
and Unit 2, respectively, and Technical Specifications (TS) for Unit 1
and Unit 2. The licensee proposes to delete obsolete and/or expired
license conditions from the Unit 1 and Unit 2 OLs, and make editorial
changes to the Unit 1 and Unit 2 OLs. Administrative changes to
specific TS for Unit 1 and Unit 2 are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
?1. Does the proposed change involve a significant
increase in the probability of occurrence or consequences of an
accident previously evaluated?
Response: No.
The proposed deletion of obsolete and/or expired license
conditions from the Unit 1 and Unit 2 OLs is administrative in
nature. The deletion of these license conditions has no impact on
plant operations since these requirements are no longer applicable.
The proposed TS changes, the renumbering of the Unit 2 OL pages, and
the correction of a typographical error in the Unit 1 OL are also
administrative in nature and do not impact CNP's current design and
licensing basis. Since the proposed changes are administrative and
do not impact plant operations or design, the changes do not involve
any significant increase in the probability or the consequences of
any accident or malfunction of equipment important to safety
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed deletion of obsolete and/or expired license
conditions from the Unit 1 and Unit 2 OLs is administrative in
nature. The proposed TS changes, the renumbering of the Unit 2 OL
pages, and the correction of a typographical error in the Unit 1 OL
are also administrative in nature. These proposed changes do not
impact plant operations or plant equipment in any manner or involve
a physical alteration to the plant, nor a change in the methods used
to respond to plant transients that has not been previously
analyzed. No new or different equipment is being installed and no
installed equipment is being removed or operated in a different
manner. Consequently, no new failure modes are introduced and the
proposed administrative changes to the Unit 1 and Unit 2 OL do not
create the possibility of a new or different kind of accident or
malfunction of equipment important to safety from any previously
evaluated. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed deletion of obsolete and/or expired license
conditions from the Unit 1 and Unit 2 OLs does not affect alarm or
trip setpoints. The proposed TS changes, the renumbering of the Unit
2 OL pages, and the correction of a typographical error in the Unit
1 OL are administrative in nature and do not impact the condition,
design, or performance of any plant structure, system or component.
Thus, the results of the accident analyses will not be affected as
any input assumptions are protected. The format changes improve
readability and appearance and do not alter any requirements. Thus,
the proposed changes do not involve a significant reduction in a
margin of safety.
In summary, based upon the above evaluation, [Indiana Michigan
Power Company] I&M has concluded that the proposed changes involve
no significant hazards consideration under the standards set forth
in 10 CFR 50.92(c), and, accordingly, a finding of ``no significant
hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 30, 2002.
Description of amendment requests: The proposed amendments would
revise the reactor trip system (RTS) and engineered safety features
actuation system (ESFAS) Technical Specification (TS) Surveillance
Requirements in TS 3/4.3.1 and TS 3/4.3.2, respectively, by increasing
(1) the channel operational test surveillance intervals for analog
channels, logic cabinets, and reactor trip breakers (RTBs), and (2) the
completion time (CT) and bypass time (BT) for the RTBs in accordance
with the evaluation and justifications presented in the referenced
document, WCAP-15376, Revision 0, ``Risk-Informed Assessment of the RTS
and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and
Completion Times,'' dated October 2000. Additionally, the proposed
amendments would remove Mode 2 applicability for the RTS low
pressurizer pressure and high pressurizer water level trips and to add
a note to TS Table 4.3-1 clarifying that channel functional testing
requirements for the reactor trip bypass breakers are only applicable
when they are racked in
[[Page 63696]]
and closed for bypassing an RTB. The proposed amendments would also
make format and capitalization changes to the affected TS pages that
improve the appearance of the TS pages, but do not affect any
requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed changes to the STIs [surveillance test intervals]
and RTB CT and BT reduce the potential for inadvertent reactor trips
and spurious actuations, and therefore do not increase the
probability of any accident previously evaluated. The proposed
changes do not change the response of the plant to any accidents and
have an insignificant impact on the reliability of the RTS and ESFAS
signals. These changes satisfy the acceptance criteria specified in
the NRC's regulatory guidance for evaluating risk-informed changes
in RG 1.174 [``An Approach for Using Probabilistic Risk Assessment
In Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis,'' dated July 1998] and RG 1.177 [``An Approach for
Plant-Specific Risk-Informed Decisionmaking: Technical
Specifications,'' dated August 1998]. The RTS and ESFAS will
continue to perform their functions with high reliability as
originally assumed in the safety analysis, and the increase in risk
is within the acceptance criteria of existing regulatory guidance;
therefore, there will not be a significant increase in the
consequences of any accidents.
The RTS and ESFAS are not accident initiators or precursors in
the safety analysis. No new initiators are created by this activity.
The proposed changes do not change any RTS or ESFAS setpoints, nor
do they alter the accident mitigation function of any system,
structure or component, design assumptions, conditions or
configuration of the facility, or the manner in which the plant is
operated and maintained. The proposed changes do not affect the
source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Further, the proposed changes do not
increase the types or amounts of radioactive effluent that may be
released offsite, nor significantly increase individual or
cumulative occupational/public radiation exposures.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
No new system interfaces or interactions are created. The
proposed changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. The proposed
changes do not result in a change in the manner in which the RTS and
ESFAS provide plant protection. The RTS and ESFAS will continue to
have the same setpoints after the proposed changes are implemented.
The proposed changes to STI, CT, and BT do not change any existing
accident scenarios, do not alter assumptions made in the safety
analysis, nor create any new or different accident scenarios.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. Redundant RTS and ESFAS trains
are maintained, and diversity with regard to the signals that
provide reactor trip and engineered safety features actuation is
also maintained. All signals credited as primary or secondary, and
all operator actions credited in the accident analyses will remain
the same. The proposed changes will not result in plant operation in
a configuration outside the design basis. The calculated impact on
risk is insignificant and meets the acceptance criteria contained in
RG 1.174 and RG 1.177.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: March 1, 2002.
Description of amendment request: The proposed amendment would
revise the containment spray nozzle inspection frequency contained in
Technical Specification Surveillance Requirement (SR) 3.6.6.9.
Specifically, the inspection frequency would be conducted ``[f]ollowing
maintenance which could result in nozzle blockage,'' rather than at the
currently specified 10-year frequency. Maintenance which could result
in nozzle blockage is controlled by procedures which establish foreign
material exclusion (FME) controls. The FME controls require post-
maintenance verification of system cleanliness and freedom from foreign
materials.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following evaluation supports the finding that operation of
the facility in accordance with the proposed change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change revises the surveillance frequency for
containment spray nozzle inspections from every ten years to
following maintenance which could result in nozzle blockage.
Analyzed events are initiated by the failure of plant structures,
systems or components. The containment spray system is not
considered as an initiator of any analyzed event. The proposed
change does not have a detrimental impact on the integrity of any
plant structure, system or component that initiates an analyzed
event. The proposed change will not alter the operation of, or
otherwise increase the failure probability of any plant equipment
that initiates an analyzed accident. As a result, the probability of
any accident previously evaluated, is not significantly increased.
This change does not affect the plant design. Due to the plant
design, the spray headers are maintained dry at the level of the
nozzles. Formation of corrosion products is unlikely due to the
corrosion resistant materials used in spray header construction. Due
to their location at the top of the containment, introduction of
foreign material from sources external to the spray nozzles is
unlikely. Since loss of foreign material control when working within
the affected boundary is the most likely cause for obstruction,
testing or inspection following such an occurrence would verify
nozzle condition, and the system would be capable of performing its
safety function. As a result, the consequences of any accident
previously evaluated are not significantly affected.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operation. No new or different type of equipment will be installed.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The margin of safety for this system is based on the capacity of
the spray headers. Since the system is not susceptible to corrosion
induced obstruction or obstruction
[[Page 63697]]
from sources external to the spray nozzles, and performance of
maintenance on the system would require evaluation of the potential
for nozzle blockage and the possible need for a test or inspection,
the likelihood that the spray nozzles might be blocked would not be
affected by the reduction in surveillance frequency. Therefore, the
capacity of the system would remain unaffected. Hence, this change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: L. Raghavan.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: July 31, 2002.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS). The proposed amendments
represent a full implementation of an alternative source term (AST) for
the Units 1, 2, and 3 operating licenses. The amendments adopt the AST
methodology by revising the current accident source term and replacing
it with an accident source term as prescribed in 10 CFR 50.67.
The AST analyses were performed using the guidance provided by
Regulatory Guide 1.183, ``Alternative Source Terms for Evaluating
Design Basis Accidents at Nuclear Power Reactors,'' dated July 2000,
and Standard Review Plan Section 15.0.1, ``Radiological Consequences
Analyses Using Alternative Source Terms.'' The four limiting design
basis accidents (DBAs) considered were the Control Rod Drop Accident,
the Refueling Accident, the Loss of Coolant Accident, and the Main
Steam Line Break Accident.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The AST and those plant systems affected by implementing AST do
not initiate DBAs. The AST does not affect the design or operation
of the facility; rather, once the occurrence of an accident has been
postulated, the new source term is an input to evaluate the
consequences. The implementation of the AST has been evaluated in
the analyses for the limiting DBAs at BFN. The equipment affected by
the proposed change is mitigative in nature and relied upon
following an accident. The proposed changes to the TS do revise
certain performance requirements. However, these changes will not
involve a revision to the parameters or conditions that could
contribute to the initiation of a design basis accident discussed in
Chapter 14 of the BFN Updated Final Safety Analysis Report.
Plant specific radiological analyses have been performed and,
based on the results of these analyses, it has been demonstrated
that the dose consequences of the limiting events considered in the
analyses are within the regulatory guidance provided by the NRC for
use with the AST. This guidance is presented in 10 CFR 50.67,
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1.
Therefore, the proposed amendment does not result in a significant
increase in the consequences or a significant increase in the
probability of any previously evaluated accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Implementation of AST does not alter any design basis accident
initiators. These changes do not affect the design function or mode
of operations of systems, structures, or components in the facility
prior to a postulated accident. Since systems, structures, and
components are operated essentially no differently after the AST
implementation, no new failure modes are created by this proposed
change. Therefore, the proposed license amendments will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The changes proposed are associated with a revision to the
licensing basis for BFN. The results of accident analyses revised in
support of the proposed change are subject to the acceptance
criteria in 10 CFR 50.67. The analyzed events have been carefully
selected, and the analyses supporting this submittal have been
performed using approved methodologies. The dose consequences of
these limiting events are within the acceptance criteria provided by
the regulatory guidance as presented in 10 CFR 50.67, Regulatory
Guide 1.183, and SRP 15.0.1.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
changes are considered to not result in a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and based on
this review, it appears that the three standards are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: August 1, 2002.
Description of amendment request: The proposed amendment would
revise the Browns Ferry design and licensing basis as described in
section 14.5.2.8 of the Updated Final Safety Analysis Report (UFSAR) to
eliminate consideration of a pressure regulator downscale failure event
as an abnormal operational transient.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment involves a change in transient analysis
assumptions and does not change the plant or the manner in which it
is operated. Therefore, the amendment has no affect on the
probability of an accident. The proposed amendment is based upon
upgrades and reliability improvements made to the main turbine
generator electro-hydraulic control system, which render the
analysis of a Pressure Regulator Downscale Failure event and
consideration of the associated consequences unnecessary. Therefore,
the proposed amendment does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment involves a change in transient analysis
assumptions and does not change the plant or the manner in which it
is operated. The only event affected, the Pressure Regulator Failure
Downscale transient, is of a type already considered. Therefore, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment eliminates the consideration of the
Pressure Regulator Downscale Failure event as an abnormal
[[Page 63698]]
operational transient based on the low likelihood of occurrence of
such an event due to improvements in the system design of the main
turbine electro-hydraulic control system. Other abnormal operational
pressurization transients as described in the UFSAR will continue to
be analyzed and ensure required margins of safety to fuel thermal
limits are maintained. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety. In
conclusion, the proposed amendment does not adversely affect the
public health and safety, and does not involve any significant
safety hazards.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 7902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: September 3, 2002.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a missed
surveillance. The delay period would be extended from the current limit
of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the
limit of the specified Frequency, whichever is greater.'' In addition,
the following requirement would be added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater than
24 hours and the risk impact shall be managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC in its application dated September
3, 2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of amendment request: August 20, 2002.
Description of amendment request: The proposed amendment would
revise Technical specifications (TSs) Table 3.3.6.1-1, Function 5.a,
Reactor Water Cleanup (RWCU) System Isolation, Main Steam Valve Vault
Area Temperature--High, to extend the frequency of the channel
calibration Surveillance Requirement (SR) from 122 days to 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed amendment changes the channel calibration
surveillance frequency from 122 days to 24 months. Under certain
circumstances, TS SR would allow a maximum surveillance interval of
30 months for the SR. An instrumentation calculation in
[[Page 63699]]
accordance with the guidelines of Generic Letter 91-04 has shown
that the reliability of protective instrumentation will be preserved
for the maximum allowable surveillance interval. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of new or
different kind of accident from any accident previously evaluated?
No. The proposed change simply extends the channel calibration
interval of instrumentation from 122 days to 24 months and does not
affect plant modes of operation. Hence, the change does not create
the possibility of any new failure mechanisms. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed amendment changes the instrument channel
calibration surveillance interval from 122 days to 24 months. An
instrumentation calculation in accordance with the guidelines of
Generic Letter 91-04 has shown safety margins are preserved with the
extended surveillance interval and that the TS allowable values are
not changed. Therefore, it is concluded that the proposed amendment
does not involve a significant reduction in a margin of safety.
The proposed amendment changes the instrument channel
calibration surveillance internal from 122 days to 24 months. An
instrumentation calculation in accordance with the guidelines of
Generic Letter 91-04 has shown safety margins are preserved with the
extended surveillance interval and that the TS allowable values are
changed. Therefore, it is concluded that the Proposed amendment does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: September 3, 2002.
Description of amendment request: The proposed amendment would
revise technical specifications Surveillance Requirement (SR) 3.0.3 to
extend the delay period, before entering a Limiting Condition for
Operation, following a missed surveillance. The delay period would be
extended from the current limit of up to 24 hours, or up to the limit
of the surveillance frequency interval, whichever is ``less,'' to up to
24 hours, or up to the limit of the surveillance frequency interval,
whichever is ``greater.'' In addition, the following requirement would
be added to SR 3.0.3: ``A risk evaluation shall be performed for any
Surveillance delayed greater than 24 hours and the risk impact shall be
managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). TVA reviewed the following
proposed NSHC determination published in the Federal Register as part
of the Consolidated Line Item Improvement Process for Technical
Specification Task Force item 358, and concluded in its application of
September 3, 2002, that the proposed NSHC determination applied to
Watts Bar.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
[[Page 63700]]
NRC Section Chief: Allen G. Howe.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Nuclear Management Company, LLC, Docket No. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin
Date of amendment request: April 30, 2002.
Brief description of amendment request: The proposed amendment
would increase the licensed reactor core power level by 1.4 percent
from 1518.5 MWt to 1540 MWt.
Date of publication of individual notice in Federal Register:
September 11, 2002 (67 FR 57630).
Expiration date of individual notice: October 11, 2002.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: August 1, 2001, and as
supplemented by letters dated June 19 and September 9, 2002.
Brief description of amendment: The amendment revises technical
specification requirements that have been superseded based on the
licensed operator training program being accredited by the Institute of
Nuclear Power Operations, promulgation of the revised 10 CFR Part 55,
and adoption of a systems approach to training as required by 10 CFR
50.120.
Date of issuance: September 24, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 154.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55009). The supplemental letters dated June 19 and September 9, 2002,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2002.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: June 26, 2002, as supplemented
on August 1, 2002.
Brief description of amendment: The amendment revised the safety
limit minimum critical power ratio values for Cycle 19 in Section 2.1.A
of the Technical Specifications, and made several editorial or
administrative corrections.
Date of Issuance: September 26, 2002.
Effective date: September 26, 2002, and shall be implemented before
Cycle 19 startup.
Amendment No.: 233.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50949). The August 1, 2002, letter provided clarifying information
within the scope of the original application and did not change the
staff's initial proposed no significant hazards consideration
determination. The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated September 26, 2002.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: August 1, 2001 as supplemented
by letters dated June 19, July 19, and September 9, 2002.
Brief description of amendment: The amendment revised, clarified,
and deleted, as appropriate, requirements regarding Facility Staff
Qualifications and licensed operator and non-licensed personnel
training programs. The changes revised requirements that have been
superseded based on licensed operator training programs being
accredited by the Institute for Nuclear Power Operations, promulgation
of the revised 10 CFR Part 55, ``Operator's Licenses,'' which became
effective on May 26, 1987, and adoption of a systems
[[Page 63701]]
approach to training as required by 10 CFR 50.120, ``Training and
qualification of nuclear power plant personnel.''
Date of issuance: September 23, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 241.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55009). Exelon's June 19, July 19, and September 9, 2002, letters
provided clarifying information within the scope of the original
application and did not change the NRC staff's proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 23, 2002.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 23, 2002, as supplemented
July 16, 2002.
Brief description of amendment: The amendment eliminates the
requirement to perform response time testing for two reactor protection
system functions and two primary containment isolation functions.
Date of issuance: October 2, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 151.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42818). The July 16, 2002, supplemental letter provided additional
clarifying information that was within the scope of the original
application and did not change the Nuclear Regulatory Commission
staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423 Millstone Power Station, Unit Nos. 1, 2, and 3 New London County,
Connecticut
Date of application for amendment: November 8, 2001, as
supplemented August 14, 2002.
Brief description of amendment: The amendments incorporate
administrative and editorial changes into the Millstone Unit No. 1
Permanently Defueled Technical Specifications (PDTS) and the Millstone
Unit Nos. 2 and 3 Technical Specifications (TSs).
Date of issuance: September 17, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 111, 270 and 212.
Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: This
amendment revises the Unit No. 1 PDTS and the Units 2 and 3 TSs.
Date of initial notice in Federal Register: December 12, 2001 (66
FR 64290). The August 14, 2002, letter provided clarifying information
that did not change the scope of the proposed action or the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 7, 2001, as
supplemented by letter dated August 7, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications 5.6.5.a by adding a few parameter limits
currently included in the Core Operating Limits Report.
Date of issuance: October 1, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 202 and 195.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2002 (67 FR
54680). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 1, 2002.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 7, 2001, as
supplemented by letter dated August 7, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications 5.6.5.a by adding a few parameter limits
currently included in the Core Operating Limits Report.
Date of issuance: October 1, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 208 & 189.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2002 (67 FR
54680). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 1, 2002.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: January 10, 2002.
Brief description of amendment: The amendment revised the technical
specifications to extend the surveillance test interval of certain
instrument channels from the current 18 months to 24 months.
Date of issuance: September 20, 2002.
Effective date: September 20, 2002, to be implemented within 30
days from the date of issuance.
Amendment No.: 179.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50951). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 20, 2002.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 24, 2001, as supplemented by
letters dated April 22 and July 29, 2002.
Brief description of amendment: The amendment extends the allowed
outage time for a Division I or Division II
[[Page 63702]]
Emergency Diesel Generator (EDG) from 72 hours to 14 days. The changes
are intended to provide flexibility in scheduling EDG maintenance
activities, reduce refueling outage duration, and improve EDG
availability during plant shutdowns.
Date of issuance: September 25, 2002.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 125.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 12, 2001 (66
FR 64292). The April 22 and July 29, 2002, supplemental letters
provided clarifying information that did not change the scope of the
original Federal Register notice or the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 25, 2002.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: January 31, 2002, as supplemented by
letter dated September 9, 2002.
Brief description of amendment: The amendment changed
administrative Technical Specification 5.5.16 regarding the Containment
Integrated Leak Rate Testing (ILRT) to allow a one-time extension of
the interval (to 15 years) for performance of the next ILRT.
Date of issuance: September 24, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 219.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7417). The September 09, 2002, supplemental letter provided
clarifying information that did not change the scope of the original
Federal Register notice or the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: August 1, 2001, as supplemented
by letters dated June 19 and September 9, 2002.
Brief description of amendments: The amendments would revise
requirements that have been superseded based on licensed operator
training programs being accredited by the Institute for Nuclear Power
Operations, promulgation of the revised 10 CFR Part 55, ``Operators''
Licenses,'' which became effective on May 26, 1987, and adoption of a
systems approach to training as required by 10 CFR 50.120, ``Training
and qualification of nuclear power plant personnel.''
Date of issuance: September 24, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 130 and 125.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55018). The supplements dated June 19 and September 09, 2002, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 24, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: April 15, 2002, as supplemented
July 8, 2002.
Brief description of amendments: The amendments change Technical
Specification surveillance requirements and allowable values for
reactor vessel steam dome pressure--high instrumentation to reflect
replacement of pressure switches with analog units.
Date of issuance: October 2, 2002.
Effective date: As of the date of issuance and shall be implemented
within 90 days for Unit 3 and prior to startup from the next refueling
outage for Unit 2.
Amendment Nos.: 195 & 188.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36930). The supplement dated July 8, 2002, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: May 14, 2002, as supplemented
by letter dated September 5, 2002.
Brief description of amendments: These amendments revised technical
specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. A TS Bases Control Program is added to
the TSs. Additionally, two administrative changes affecting TS Section
6.2.2, ``Unit Staff,'' and Section 6.5.1.2, ``Composition,'' were
incorporated.
Date of issuance: October 2, 2002.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 162 and 124.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 27, 2002 (67 FR
55041). The supplement dated September 5, 2002, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register on August 27, 2002 (67 FR 55041).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
[[Page 63703]]
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
277, Peach Bottom Atomic Power Station, Unit 2, York County,
Pennsylvania
Date of application for amendment: June 10, 2002, as supplemented
August 2, 2002.
Brief description of amendment: This amendment revises the
Technical Specifications (TSs) for the safety limit for the minimum
critical power ratio from its current value of 1.09 to 1.07 for two
recirculation-loop operation, and from 1.10 to 1.09 for single
recirculation-loop operation.
Date of issuance: September 23, 2002.
Effective date: As of the date of issuance, to be implemented prior
to startup for cycle 15 operations, scheduled for September 2002.
Amendment No.: 246.
Facility Operating License No. DPR-44: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50953). The August 2, 2002, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 23, 2002.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 21, 2002.
Brief description of amendment: The amendment modifies the
Technical Specification Surveillance Requirement 3.7.3.1 to improve
consistency with Cooper Nuclear Station (CNS) Amendment No. 185,
approved on March 13, 2001, and eliminate unnecessary restrictions
regarding how the reactor equipment cooling system surge tank level is
monitored.
Date of issuance: September 18, 2002.
Effective date: September 18, 2002.
Amendment No.: 194
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15624). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 18, 2002.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: June 24, 2002.
Brief description of amendment: The amendment revised Surveillance
Requirement 3.0.3 to extend the delay period, before entering a
Limiting Condition for Operation, following a missed surveillance. The
delay period was extended from the current limit of ``* * * up to 24
hours or up to the limit of the specified Frequency, whichever is
less'' to ``* * *up to 24 hours or up to the limit of the specified
Frequency, whichever is greater.'' In addition, the following
requirement was added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
Date of issuance: October 2, 2002.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 107.
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53987). The staff's related evaluation of the amendment is contained in
a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: December 19, 2001, as
supplemented April 19, 2002.
Brief description of amendment: The amendment extends the time for
completing required action A.1 of TS 3.8.4, ``Electrical Power
Systems--DC Sources--Operating,'' for restoring the 125 volt direct
current (VDC) electrical power subsystem to operable status. The
change, in effect, provides for replacement of 125 VDC batteries 1D1
and 1D2 while the plant is at power. The time is extended on a one-time
basis, and for each battery division separately, from 8 hours to 10
days. The one-time change also requires that required features be
declared inoperable when the associated 125 VDC source is inoperable
and the redundant required features are also inoperable for at least 4
hours.
Date of issuance: October 1, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 247.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 5, 2002 (67 FR
5329). The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 1, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: February 8, 2002, as
supplemented June 21, 2002.
Brief description of amendment: Amendment changes Technical
Specification 5.0 to be consistent with Technical Specifications Task
Force Change No. 258, Revision 4, ``Changes to Section 5.0,
Administrative Controls.''
Date of issuance: October 2, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 248.
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15625). The supplemental letter contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: June 7, 2002, as supplemented
August 20 and 29, 2002.
Brief description of amendment: The amendment would revise the
Kewaunee Nuclear Power Plant Technical Specification (TS) Sections for
administrative changes:
(1) Section 1-``Definitions,'' (2) Section 2-``Safety Limits and
Limiting Safety System Settings,'' (3) Section 5-``Design Features,''
and (4) Section 6-``Administrative Controls.''
[[Page 63704]]
The administrative changes include capitalizing defined words,
formatting section titles, renumbering pages and correcting
miscellaneous grammar and punctuation errors.
Date of issuance: September 19, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 162.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 23, 2002 (67 FR
48220). The supplemental letter contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 19, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 7, 2002, as supplemented
July 19 and September 11, 2002.
Brief description of amendment: The amendment revised the Kewaunee
Nuclear Power Plant technical specification (TS) requirements for
meeting surveillances in TS 4.0.a, TS requirements for missed
surveillances in TS 4.0.c, and TS requirements for a Bases control
program consistent with TS Bases Control Program described in Section
5.5 of NUREG-1431, Standard TS for Westinghouse Plants, Revision 2.
Date of issuance: September 24, 2002.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 163.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42829). The supplements dated July 19 and September 11, 2002, provided
clarifying information that did not change the scope of the May 7,
2002, application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: December 21, 2001, as
supplemented April 26, 2002.
Brief description of amendment: The amendment revises Technical
Specifiction (TS) Sections 3.7/4.7, ``Containment Systems,'' to (1)
clarify existing requirements, (2) make editorial changes, (3) revise
limiting conditions for operation (LCOs) and surveillance requirements,
and (4) add certain LCOs.
Date of issuance: September 23, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 130.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34490). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 23, 2002.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: October 17, 2001, as
supplemented June 25, 2002.
Brief description of amendment: The amendment revises the
multiplier values for the single-loop operation average planar linear
heat generation rate to account for the use of General Electric (GE)14
fuel.
Date of issuance: October 2, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 131.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 14, 2001 (66
FR 57122). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2002.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of application for amendments: September 13, 2001, as
supplemented by letter dated February 27, 2002.
Brief description of amendments: The amendment revises Technical
Specification (TS) 3.7.16, ``Spent Fuel Pool Boron Concentration''; TS
3.7.17, ``Spent Fuel Assembly Storage--Region 1/Region 2''; and TS 4.3,
``Fuel Storage,'' for Diablo Canyon Nuclear Power Plant Units 1 and 2,
to allow the use of credit for soluble boron in the spent fuel pool
criticality analysis.
Date of issuance: September 25, 2002.
Effective date: September 25, 2002, and shall be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-154; Unit 2-154.
Facility Operating License Nos. DPR-80 and DPR-82: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55020). The February 27, 2002, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 25, 2002.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M.
Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendment request: March 4, 2002, as supplemented by letter
dated July 11, 2002.
Brief Description of amendment: The proposed amendment revises
Technical Specifications (TS) 5.5.9.3.a, ``Steam Generator Tube
Surveillance Program, Inspection Frequencies.'' Specifically, the
proposed changes revise the Farley Nuclear Plant, Unit 1 TS to allow a
40-month inspection interval after its first (post-replacement)
inservice inspection, rather than after two consecutive inspections
resulting in C-1 classification.
Date of issuance: September 20, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 157.
Facility Operating License No. NPF-2: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53991). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 20, 2002.
[[Page 63705]]
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: September 20, 2001, as
supplemented by letters dated January 24, April 25, July 3, and July
16, 2002.
Brief description of amendments: The amendments revise the
Technical Specifications to support extension of certain surveillance
requirements from ``92 days'' to ``92 days on an alternate test
basis.''
Date of issuance: September 26, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 234 and 176.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 28, 2001 (66
FR 59514). The supplements dated January 24, April 25, July 3, and July
16, 2002, provided clarifying information that did not change the scope
of the September 20, 2001, application nor the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 26, 2002.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 2, 2001, as supplemented by
letters dated March 6, April 2, and June 25, 2002. The supplemental
information provided clarification that did not change the scope or the
initial no significant hazards consideration determination.
Brief description of amendments: The amendments revise the
Technical Specification permitting a one time extension of Title 10 of
the Code of Federal Regulations, Part 50, Appendix J, Option B,
Performance-Based Leakage-Test Requirements.
Date of issuance: September 17, 2002.
Effective date: September 17, 2002.
Amendment Nos.: Unit 1-143; Unit 2-131.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50959). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 17, 2002.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 4, 2002 (TSC 00-04).
Description of amendment request: The amendments relocated certain
Technical Specification (TS) surveillance requirements to the Sequoyah
Technical Requirements Manual.
Date of issuance: September 5, 2002.
Effective date: September 5, 2002.
Amendment Nos: 277 and 268.
Facility Operating License No. DPR-79: Amendments revise the TSs.
Date of initial notice in Federal Register:April 16, 2002 (67 FR
18648). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 5, 2002.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 21, 2001, as
supplemented by letters dated June 11, July 19, August 9 and 30, and
September 5 and 12, 2002 (TS 00-06).
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to allow irradiation of up to 2256
tritium-producing burnable absorber rods.
Date of issuance: September 30, 2002.
Effective date: As of the date of issuance and shall be implemented
prior to irradition of TPBARs.
Amendment Nos.: Unit 1-278, Unit 2-269.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the TSs.
Date of initial notice in Federal Register: December 17, 2001 (66
FR 65000). The supplemental letters provided clarifying information
that did not expand the application beyond the scope of the initial
notice and did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in an Environmental Assessment dated September 23, 2002 (67 FR 59581)
and in a Safety Evaluation dated September 30, 2002.
No significant hazards consideration comments received: Comments
were received in response to the staff's proposed no significant
hazards consideration determination that was published in the December
17, 2001, Federal Register, from Dr. Kenneth D. Bergeron and The Blue
Ridge Environmental Defense League (BREDL). BREDL's comments
incorporated Mr. Bergeron's comments by reference. These comments were
addressed by the staff in a letter from Dr. Brian Sheron to Mr.
Bergeron dated September 6, 2002, with a copy to BREDL (Accession No.
ML022410310).
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 12, 2001, as
supplemented September 17, 2002 (TS 01-04).
Brief description of amendments: The proposed amendments would
change the Sequoyah Nuclear Plant, Units 1 and 2 Technical
Specification (TS) 3/4 6.5.1 and associated Bases to reflect an
increase in the ice condenser basket weight from 1071 pounds to 1145
pounds and the total ice condenser ice weight from 2,082,024 pounds to
2,225,880 pounds. This change is being made in response to a reanalysis
by Westinghouse Electric Company that identified a modeling input error
used in the original analysis.
Date of issuance: September 30, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 279 & 270.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: January 8, 2002 (67 FR
934). The September 17, 2002, supplement contained clarifying
information only, and did not change the initial no significant hazards
consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 30, 2002.
No significant hazards consideration comments received: No.
[[Page 63706]]
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: August 20, 2001, as supplemented
by letters of October 29, November 14, November 21, December 7,
December 19, 2001, and January 14, February 19, February 21, May 21,
May 23, and July 30, 2002.
Brief description of amendment: The amendment allows Watts Bar
Nuclear Plant, Unit 1, to irradiate up to 2304 tritium-producing
burnable absorber rods in the reactor core each fuel cycle.
Date of issuance: September 23, 2002.
Effective date: As of the date of issuance and shall be implemented
prior to starting up from the outage where TVA inserts tritium-
producing burnable absorber rods in the core.
Amendment No.: 40.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 17, 2001 (66
FR 65005). The supplemental letters provided clarifying information
that was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in an Environmental Assessment dated August 20, 2002 (ADAMS Accession
No. ML022320905) and in a Safety Evaluation dated September 23, 2002.
No significant hazards consideration comments received: Comments
were received in response to the staff's proposed no significant
hazards consideration determination (66 FR 65005) from Dr. Kenneth D.
Bergeron and The Blue Ridge Environmental Defense League (BREDL).
BREDL's comments incorporated Dr. Bergeron's comments by reference.
These comments were addressed by the staff in a letter from Dr. Brian
Sheron to Dr. Bergeron dated September 6, 2002, with a copy to BREDL
(Accession No. ML022410310). The staff made a final determination that
the amendment involves no significant hazards consideration, which is
contained in the Safety Evaluation dated September 23, 2002.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: October 23, 2001, as supplemented by
letters dated July 23, August 29, and September 6, 2002.
Brief description of amendments: The amendments revise TS 5.5.9,
``Steam Generator Tube Inspection Report,'' to permit installation of
leak-tight sleeves in the Comanche Peak Steam Electric Station, Unit 1,
steam generators as an alternative to plugging defective steam
generator tubes.
Date of issuance: September 25, 2002.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 101.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 26, 2001 (66
FR 66473). The July 23, August 29, and September 6, 2002, supplemental
letters provided clarifying information that did not change the scope
of the original Federal Register notice or the original no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 25, 2002.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 17, 2002.
Brief description of amendment: The amendment revises Limiting
Conditions for Operation (LCOs), Required Actions for LCOs,
Surveillance Requirements, and Tables specifying requirements on
instrumentation in the following Technical Specifications: (1) TS
3.3.6, ``Containment Purge Isolation Instrumentation''; (2) TS 3.3.7,
``Control Room Emergency Ventilation System (CREVS) Instrumentation'';
(3) TS 3.3.8, ``Emergency Exhaust System (EES) Actuation
Instrumentation''; and (4) TS 3.9.4, ``Containment Penetrations.'' The
revisions allow the equipment hatch and the emergency air lock to be
open in refueling outages during core alterations and/or movement of
irradiated fuel within containment.
Date of issuance: September 9, 2002.
Effective date: September 9, 2002, and shall be implemented,
including the incorporation of the changes to the Bases of the
Technical Specifications and to the Final Safety Analysis Report for
Callaway, as described in the licensee's letter of June 17, 2002, prior
to entry into Mode 6 during Refueling Outage 12 that is scheduled for
the Fall of 2002.
Amendment No.: 152.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 23, 2002 (67 FR
48222). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 9, 2002.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: February 26, 2002, as
supplemented by letter dated July 15, 2002.
Brief description of amendment: These amendments revise the
surveillance frequency of the quench and recirculation spray system
nozzles, from a time period of every 10 years to whenever maintenance
is conducted that could contribute to nozzle blockage.
Date of issuance: October 1, 2002.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 233 and 215.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments change
the Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21296). The supplemental letter dated July 15, 2002, provided
clarifying information that did not change the scope of the February
26, 2002, application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 2002.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 4th day of October 2002.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 02-25990 Filed 10-11-02; 8:45 am]
BILLING CODE 7590-01-P